This invention relates broadly to nuclear fuel elements having a composite cladding comprised of a metal barrier of sponge or crystal bar zirconium bonded to the inside surface of a zirconium alloy tube, and more particularly to such a composite cladding having improved strength and oxidation resistance.
Nuclear reactors are presently being designed, constructed and operated in which the nuclear fuel is contained in fuel elements which can have various geometric shapes, such as plates, tubes, or rods. The fuel material is usually enclosed in a corrosion-resistant, nonreactive heat-conductive container or cladding. The elements are assembled together in a lattice at fixed distances from each other in a coolant flow channel or region forming a fuel assembly, and sufficient fuel assemblies are combined to form the nuclear fission chain reacting assembly or reactor core capable of a self-sustained fission reaction. The core, in turn, is enclosed within a reactor vessel through which a coolant is passed.
The cladding serves several purposes and two primary purposes are: first, to prevent contact and chemical reactions between the nuclear fuel and the coolant or the moderator if a a moderator is present, or both if both coolant and moderator are present; and second, to prevent the radioactive fission products, some of which are gases, from being released from the fuel into the coolant or the moderator or both if both coolant and moderator are present Useful cladding materials include stainless steel, aluminum and its alloys, zirconium and its alloys, certain magnesium alloys, and others. The failure of the cladding, i.e., a loss of the leak tightness, can contaminate the coolant or moderator and the associated systems with radioactive long-lived products to a degree which interferes with plant operation.
Problems have been encountered in the manufacture and in the operation of nuclear fuel elements which employ certain metals and alloys as the clad material due to mechanical stresses or chemical reactions occurring with these cladding materials under certain circumstances. Zirconium and its alloys, under normal circumstances, are excellent nuclear fuel claddings since they have low neutron absorption cross sections and at temperatures below about 750.degree. F. (about 398.degree. C.) are strong, ductile, extremely stable and nonreactive in the presence of demineralized water or steam which are commonly used as reactor coolants and moderators.
However, fuel element performance has revealed a problem with the brittle splitting of the cladding due to the combined interactions between the nuclear fuel, the cladding and the fission products produced during nuclear fission reactions. It has been discovered that this undesirable performance is due to localized mechanical stresses on the fuel cladding resulting from differential expansion and friction between the fuel and the cladding. Fission products are created in the nuclear fuel by the fission chain reaction during operation of a nuclear reactor, and these fission products are released from the nuclear fuel and are present at the cladding surface These localized stresses and strain in the presence of specific fission products, such as iodine and cadmium, are capable of producing cladding failures by phenomena known as stress corrosion cracking and liquid metal embrittlement.
Within the confines of a sealed fuel element, hydrogen gas can be generated by the slow reaction between the cladding and residual water inside the cladding, and this hydrogen gas may build up to levels which, under certain conditions, can result in localized hydriding of the cladding, with concurrent localized deterioration in the mechanical properties of the cladding. The cladding may also be adversely affected by such gases as oxygen, nitrogen, carbon monoxide and carbon dioxide over a wide range of temperatures.
The zirconium cladding of a nuclear fuel element is exposed to one or more of the gases listed above and fission products during irradiation in a nuclear reactor, and this occurs in spite of the fact that these gases may not be present in the reactor coolant or moderator and further, may have been excluded as far as possible from the ambient atmosphere during manufacture of the cladding and the fuel element. Sintered refractory and ceramic compositions, such as uranium dioxide, and other compositions used as nuclear fuel, release measurable quantities of the aforementioned gases upon heating, such as during fuel element manufacture and further release fission products during nuclear fission chain reactions. Particulate refractory and ceramic compositions, such as uranium dioxide powder and other powders used as nuclear fuel, have been known to release even larger quantities of the aforementioned gases during irradiation. These released gases are capable of reacting with the zirconium cladding containing the nuclear fuel.
Thus, in light of the foregoing, it has been found desirable to minimize attack on the zirconium cladding from water, water vapor and gases, especially hydrogen, reactive with the cladding from inside the fuel element throughout the time the fuel element is used in the operation of nuclear power plants.
Two particularly effective approaches for inhibiting the degradation of zirconium and zirconium alloy nuclear fuel cladding tubes are described in U.S. Pat. Nos. 4,200,492 and 4,372,817, the disclosures of which are incorporated herein by reference. A composite cladding tube described therein comprises a barrier layer of either high purity zirconium (such as crystal bar zirconium) or moderate purity zirconium (such as sponge zirconium) metallurgically bonded on the inside surface of a zirconium alloy tube. The composite cladding encloses the nuclear fuel material, leaving a gap between the fuel and the cladding. The barrier layer shields the alloy tube from the nuclear fuel material held in cladding as well as shielding the alloy tube from fission products and gases. The barrier layer typically has a thickness equal to about 1 to about 30 percent of the thickness of the composite cladding. In both caes, the barrier layer remains relatively soft during irradiation and minimizes localized strain inside the nuclear fuel element, thus serving to protect the alloy tube from both stress corrosion cracking and liquid metal embrittlement. The alloy tube portion of the cladding is otherwise unchanged in design and function from previous practice for a nuclear reactor and is selected from conventional cladding materials, such as zirconium alloys.
It is disclosed in U.S. Pat. Nos. 4,200,492 and 4,372,817 that the high and moderate purity zirconium metal forming the metal barrier in the composite cladding, even after prolonged irradiation, is able to maintain desirable structural properties such as yield strength and hardness at levels considerably lower than those of conventional zirconium alloys. In effect, the metal barrier does not harden as much as conventional zirconium alloys when subjected to irradiation, and this together with its initially low yield strength enables the metal barrier to deform plastically and relieve pellet-induced stresses in the fuel element during power transients. Pellet induced stresses in the fuel element can be brought about, for example, by swelling of the pellets of nuclear fuel at reactor operating temperatures (300.degree. to 350.degree. C.) so that the pellet comes into contact with the cladding.
The nuclear fuel elements described in U.S. Pat. Nos. 4,200,492 and 4,372,817 are a substantial improvement over elements which do not include internal zirconium barrier layers. Certain problems, however, are encountered in the fabrication and utilization of such nuclear fuel elements. First, the softness and large grain size of the zirconium sponge barrier layer has been found to promote surface cracking, or microfissuring, during the fabrication of the composite cladding, particularly during the stage where the tube shell is formed into the tubing. The microfissures can extend up to 10 microns into the zirconium barrier layer (which is typically about 75 microns thick), serving as initiation sites for stress corrosion cracking. Second, the relatively pure zirconium liners will oxidize rapidly if the composite cladding is breached and water or steam enters the fuel rod during operation of the reactor.
It would thus be desirable to provide an improved nuclear fuel rod of the type generally disclosed in U.S. Pat. Nos. 4,200,492 and 4,372,817, wherein the tendency of the relatively pure zirconium barrier layer to crack during fabrication and oxidize during operation are largely inhibited. It would be particularly desirable if such cracking and oxidation inhibition can be achieved without reducing the effectiveness of the zirconium barrier layer, particularly the ability of the barrier layer to deform plastically and relieve pellet-induced stresses in the fuel element during power transients.